1. 1 Class 19 Nuclear Spent Fuel Reprocessing

2. 2 Introduction During its time in the reactor, the fuel (bars, tubes…) is subject to important physical and composition modifications due to the neutrons irradiation: The fissile material content (U-235 or Pu-239, Pu-241) decreases progressively by fission U-239 generates by capture Pu-239, which disappears by capture or by fission. –The capture reactions lead to the formation of Pu-240, Pu-241, Pu- 242 (with Pu-241 fissile). This apparition of new fissile isotopes compensate only partially the diminution of the global content of fissile matter Apparition of new elements in the fuel, baneful to the chain reaction progress –Transuranic elements (Np, Am, Cm) –Fission Products (Sr, Cs, Tc…) some of them are neutrons poison such as Gd

3. 3 Introduction (2) The modifications of the fuel composition associated to the strong heat release by fission provoke important changes in the physical state of the fuel. Crystals structure modifications (holes or concentrations of atoms) Variation of the volume: –The volume occupied by the atoms created by fission is greater than the one of the disappeared matter. –Moreover, some fission products are gaseous and their solubility in uranium is practically non-existent

4. 4 Introduction (3) All of these changes will alter the physical properties and the structure of the fuel with modifications of the thermal, mechanical, dimensional characteristics. Consequently the cladding can be deteriorated, which can go to the formation of cracks or even break.

5. 5 Introduction (4) The following implies that, after a certain period of irradiation time, it is necessary to take out the fuel from the reactor. –Decrease of the content of fissile material –Progressive poisoning of the fuel –Risk of cladding break This operation is performed before the consumption of all initial fissile material

6. 6 Introduction (5) The reprocessing activity answers 2 objectives: –Recovery of the fissile material (U-239, Pu-239, Pu-241) to reuse it for the fabrication of new fuel elements (example Recycling Pu via MOX fuel fabrication, recycling of U from PWR for new enrichment) –Separation of nuclear waste (activation and fission products) as a function of their pollution in order to store then to avoid any potential danger and release towards the biosphere

7. 7 Irradiated Fuels Characteristics The irradiated fuels are taken out of the nuclear reactor after a certain time that can vary : (PWR: 3 years, fast breeder reactor: 2 years) The following table is presenting the principal characteristics of some nuclear fuels

8. 8 Principal characteristics of irradiated fuels Reactor typeFormNatureCladdingCombustion Rate (Mwd/t) UNGGBarsU metal slightly alloyed Magnesium3,000 to 5,000 Heavy water Light water Fast neutrons RodsNatural UO 2 Enriched UO 2 (< 5%) (U,Pu)O 2 with 15 to 20% of Pu Zircaloy Zircaloy,stainless steel Stainless steel 10,000 to 15,000 20,000 to 40,000 50,000 to 100,000 Submarine reactors Plates assembly U-Al U-Zr (enriched U up to 93% Aluminum Zircaloy 30% < 30%

9. 9 Irradiated Fuels Characteristics (2) If we consider only the characteristics important to the reprocessing, we can distinguish: –The composition (metal, alloy, oxide, carbide…) –The enrichment of the fissile material density –The structure or the form (bars or tubes, rods, plates, spheres..) –Combustion rate whose depend the activity , ,  and the residual calorific power after irradiation

10. 10 Irradiated Fuels Characteristics (3) For the “reprocessor”, the irradiated fuel can be considered as a mixture of 5 families of compounds: –The fissile material (U-235, Pu-239 and Pu-241), which represents the noble part of the fuel and whose recovery is the main goal of reprocessing –The fertile material (U-238, Pu-240) –Heavy isotopes, they are neither fissile nor fertile (U- 236, Np-237, Pu-242, Am and Cm) –The fission products (principal source of  and  activity) –The other metals (Mg, Al, Zr, metals coming from stainless steel) part of the fuel or forming the cladding

11. 11 Irradiated Fuels Characteristics (4) The composition of one ton of irradiated fuel coming from the following reactors –UNGG –PWR –Fast neutrons Reactors is reproduced in the next table and is illustrated for the PWR case in the following slide The content of fission products and of plutonium increase with the fuel rate as UNGG > PWR > Fast BR

12. 12 Fuel characteristics from nuclear reactors Reactor Type 1200MW UNGGPWR Fast Neutrons reactors Rod containing (U and Pu)O 2 Stainless Steel Cladding Rod containing UO 2 enriched (3.25%) Zircaloy Cladding U metallic Bar alloyed with Mo or Al Mg cladding Form and Initial Composition Average Burn up In the reactor 80,000 MWd/t for the fuel in the core reactor  and  activity Residual power After 150 days cooling 1,800 Ci = 75 W per kg of fuel 4,500 Ci = 20 W per kg of fuel 1,400 Ci = 6 W per kg of fuel Quantity To be treated Quantity of principal FP (g/ton Total FP Actinides (g/ton) Total An

13. 13 Characteristics of fuel irradiated during 3 years in a PWR Initial Fuel (1,000 kg) Irradiated Fuel (1,000 kg) 3 years Fission Products Different Isotopes Of Pu (9 kg)

14. 14 The repartition in mass of the fission products can be deduced from the curves: Yield of fission vs. mass number of FP See next slide

15. 15 Distribution curves of the FP of U-235 Yield in mass of the fission (%) Fast Neutrons (14 MeV) Thermal Neutrons Mass Number of the Fission Product

16. 16 Distribution curves of the FP of U-233, U-235, Pu-239 (Thermal Neutrons) Mass Number of the Fission Product Yield in mass of the fission (%)

17. 17 Irradiated Fuels Characteristics (5) An important characteristic of irradiated fuels is their residual calorific power, consequence of the activities  and  of fission products and activation products It decreases over time, as presented in the next table

18. 18 Residuaire Power of irradiated Fuels Residual Power in w/kg after Reactor Type Fuel Rate MWd/t 90 days 120 days150 days180 days UNGG4,0008.77.05.84.9 PWR30,00030242017 FBR50,000 to 100,000 80 110 65 90 55 80 47 72

19. 19 Irradiated Fuels Characteristics (6) The reprocessing of these different types of fuels is not so much different, nevertheless, on needs to take into account the fissile material density, the content of fission products, the calorific power produced, the structure of the fuel elements.

20. 20 The chemical Treatment Objectives and constraints The problem is as follows: Obtain separately and with a high percentage yield Uranium and Plutonium decontaminated from the fission products, in order to manipulate them later as if they were materials that have never been irradiated The decontamination factor of U and Pu in FP is between 10 7 to 10 8 The purification of U and Pu towards non radioactive elements need to be also very specific

21. 21 The chemical Treatment (2) The specifications to respect for finished products U and Pu issued from reprocessing of LWR and fast neutrons reactors fuels are described in the next table

22. 22 Constraints for recovery and purification of finished products EXTRACTION YIELD ≥ 99.5% Decontamination Factors for Uranium and Plutonium (FP) Cooled Fast Neutrons Reactor Fuels (6 m) ,  Activity ~ 5 10 6 Ci/ton U+Pu  Activity ~ 9.10 5 Ci/ton U+Pu (100,000 MWD/t) -Np 385g/ton U+Pu -Pu 10,000 g/ton U + Pu FINISHED PRODUCTS SPECIFICATIONS ,  Activity ~ 0.5 mCi/g U ~ 1 to 2 mCi/g Pu  Activity ~ 1,500 to 15,000 dpm/g U LWR 7,500 dpm = 5  g Np 7,500 dpm = 10 -8 g Pu Fast Neutrons 7,500 dpm = 5  g Np 7,500 dpm = 10 -8 g Pu Cooled LWR Fuels (3 years) ,  Activity ~ 7.5 10 5 Ci/ton U+Pu  Activity ~ 10 5 Ci/ton U+Pu ``-Np 385g/ton U+Pu -Pu 10,000 g/ton U + Pu LWR PLANT DF U ,  = 1.5.10 6 Pu = 2.10 5 Np = 100 (77) DF Pu ,  = 10 7 to 7.10 7 U ~ 6.4.10 5 (150 ppm in Pu) FBR PLANT DF U ,  = 10 7 Pu = 1.3.10 7 Np = 36 (or 360) DF Pu ,  = 4.10 7 U ~ 1.5.10 5 Np 3 to 4

23. 23 Constraints to respect during reprocessing operations are numerous. Indeed, The fuel is inside a water proof cladding, in general refractory to the usual chemical reagents The intense radioactivity implies very special work conditions The fissile material masses require to be very careful towards criticality risks The radioactivity of the effluents release into the environment need to be very small and lower than the limits established by agencies such as EPA

24. 24 The next tables show the performances to reach for the liquid and gaseous effluents. The purification factors, ratio between quantity that enters the plant and quantity that exits the plant (via effluents) are depending on the plant but are in general very high, specially for alpha emitters.

25. 25 Respect of the environment - Liquid Effluents Purity Factor* for the most radioactive Effluents La Hague Pur. Factor ,  = 2.6 10 4 Pur. Factor  = 2.6 10 4 ANNUAL ACTIVITY TREATED In UP2 800 + UP 3 (1600 t/year – LWR) SITE: La Hague, France ,  : 1.2 10 9 Ci/year 3 H : 1.12 10 6 Ci/year  : 1.6 10 8 Ci/year RELEASE ASKED FOR La Hague ,  : 45,000 Ci/year 3 H : 5.3 10 6 Ci/year  : 90 Ci/year The purification factors, ratio between quantity that enters the plant and quantity that exits the plant (via effluents) NB: American norms for liquid release  = 0.5 10 -3 Ci per GWe and per year for a PWR cycle. This norm implies a purity factor of 10 10 for the high active liquid effluents

26. 26 Respect of the environment Gaseous Effluents Purity Factor* for the most radioactive Effluents La Hague Pur. Factor Kr = 1 Pur. Factor  > 30 Pur. Factor , ,  = 10 8 ANNUAL ACTIVITY TREATED In UP2 800 + UP 3 (1600 t/year – LWR) SITE: La Hague, France Kr  = 1.7 10 7 Ci/year 3 H = 1.1 10 5 Ci/year  = 61 Ci/year Others FP = 10 9 Ci/year ATMOSPHERIC RELEASE ASKED FOR La Hague Kr-85 : 10 7 Ci/year 3 H : 5 10 4 Ci/year I-129 and I-131 : 2Ci/year Others FP = 10 Ci/year NB: American norms for gasrelease Kr-85 = 50,000 Ci: I = 5.10 - 3 Ci per GWe and per year for LWR which implies a Pur. Factor Iode > 200 and Pur. Factor for Kr = 10 for La Hague

27. 27 Principle The principle of chemical reprocessing of nuclear fuels relies essentially on liquid- liquid extraction. This choice implies the dissolution of irradiated fuels in an aqueous solution, after elimination of cladding material, followed by the realization of the liquid-liquid extraction cycles that leads to the chemical separation –Uranium + Plutonium towards fission products and other metals –Uranium from Plutonium

28. 28 The universal process used today is the PUREX Process. It uses a nitric dissolution of the fuels A specific separation of U and Pu by extraction, using a solvent n-Tri- butylphosphate diluted in an aliphatic diluent (dodecane). The scheme of the PUREX Process is reproduced next slide

29. 29 Liquid effluents Unloading Off-Gas Treatment Iodine Kr-Xe Storage GAS TREATMENT VENTILATION Storage Pool shearing Dissolution Clarification 1st cycle TBP Extraction 2nd cycle 3rd cycle 2nd cycle 3rd cycle Pu nitrate U nitrate PuO 2 UO 3 UO 2 UF 4 Pu U HLW FP, Np, Cm, Am Concentration Denitration Interim liquid storage Interim Storage of glass blocks in well Interim Storage under water TREATMENT SLUDGE Gaseous effluents Solid compounds Liquid Effluents SOLID WASTES Oxides Fuel Assemblies WASTES Vitrification U(VI)-Pu(IV) UF 6 RE-INTRODUCTION IN THE FUEL CYCLE

30. 30 The different steps of the Process Deactivation and transport to the plant The first destination of the irradiated fuel is close to the reactor This is the storage pool where the assemblies are stored under a few meters of water This period of cooling allow an important fraction of the radioactivity to cool down After a few months, the most instable radionuclides, whose the half life time is in minute, hour, day, have practically disappeared and there are only the fission products left with long half-life time. It is then unnecessary to wait any longer, the irradiated fuel can be transported towards the reprocessing plant.

31. 31 Transport The fuel assemblies consisting of rod bundles measuring about 4 m long, holding usually 264 rods (PWR 17*17) initially containing UO 2 based on uranium enriched to about 3.5% U-235, irradiated to about 2.85*10 12 J/kg (33,000MWd/t) for PWR fuel, ( and about 2.42*10 12 J/kg (28,000 MWD/t for BWR fuels) are transported in shielded casks from the power plant sites to the reprocessing plants. These casks, designed for the simultaneous transport of several fuel assemblies are very heavy (about 100 t) and complex machines that must guarantee transport safety.

32. 32 Transport (2) On their arrival at the reprocessing plant, the fuels are unloaded form the transport casks. This delicate operation is normally performed after the loaded cask is immersed in water To simplify this unloading operation, AREVA has successfully installed a dry unloading facility at La Hague, an operation designed to shorten unloading time and to minimize the volume of contaminated effluents to be treated. After unloading, the fuel assemblies are stored under water in pools awaiting reprocessing

33. 33 Spent fuel storage capacity The spent oxide fuel storage capacity pf the reprocessing pants is already considerable. As an example, La Hague has 5 pools with a total capacity of 10300 t currently in operation.

34. 34 The different steps of the Process (2) Decladding and dissolution clarification First decladding takes place Decladding can be performed –Chemically if the material of the cladding and the form of the fuel allow it (case of the magnesium cladding for UNGG) –Mechanically for the UNGG, PWR and fast neutrons reactors fuels

35. 35 Decladding operations (La Hague, Windscale … Thumb Wheel Cylinder Rod end

36. 36 Shearing

37. 37 Shearing To enable nitric acid solution to attack the fuel, it is necessary to chop the assemblies. This operation is performed at the reprocessing plant, after cutting the top and bottom ends of the assemblies, using a horizontal shear which accommodates the complete bundle. Each rod of the fuel element is thus broken into pieces about 30X 35 10 -3 m long, which contain all or part of the nuclear material. At this stage a fraction of the gaseous fission products escapes through the dissolver off gas where it is subjected to iodine trapping by NaOH scrubbing and absorption on silver loaded inorganic solid sorbent.

38. 38 The different steps of the Process (3) Once the decladding, and shearing are performed, the fuel dissolution can take place. It is performed on pieces of the cladding and the fuel by mixing to a nitric acid solution The dissolution reactions are: and UNGG PWR and Fast Neutrons Reactors

39. 39 Dissolution

40. 40 Small amounts of residues remain during the dissolution phase of PWR and fast neutrons reactor fuels. These residues are made of small part of cladding and polymetallic inclusions containing fission products: Mo, Ru, Rh, Te, Pd + U, Pu whose the quantity increases with the combustion rate. Dissolution (2)

41. 41 Fuel dissolution produces a solution with the following approximate composition: U(VI) = 250 kg/m 3 (g/L) Pu(IV) = 2.5 kg/m 3 (g/L) FP = 9 kg/m 3 (g/L) The  and  activities are approximately 200 Ci The  activity is approximately 1.87 Ci/L Dissolution (3)

42. 42 The fuel is not totally dissolved and about 3 kg of FP is not dissolved. These are the highly active dissolution fines consisting of FP: Pd, Rh, Ru and Mo. A small fraction of Pu is found trapped in these residues. Dissolution (4)

43. 43 Implementation of the dissolution operations must take into account the criticality risks. Tow alternatives are available: Poisoning of the dissolution liquor by a neutron absorbent (such as gadolinium) and Criticality safety guaranteed by the geometry. The next figure presents 2 types of continuous dissolver, (rotary dissolver UP3, La Hague and helicoidal continuous dissolver (TOR, Cadarache). Dissolution (5)

44. 44 Continuous Rotary Dissolver Metal Feed Wheel = 4m engine Discharge of Insolubles Dissolution Solution Nitric Acid Feed Fast Annular Continuous Dissolver Feed Dissolution Solution Solid Wastes engine Suspension with spring

45. 45 Clarification Before being sent to the U and Pu extraction cycles, the dissolution liquors must be clarified to remove any traces of solid particles, whose presence would be detrimental to the extraction operations. The solution is cleared by centrifugation or filtration and adjusted to 4M HNO 3 before going towards the extraction unit.

46. 46 The gas emitted from the dissolver contain nitrous vapors NOx and certain fission products, the noble gas Krypton and Xenon as well as Iodine (volatile as I 2 form). Iodine is trapped as stable form PbI 2 and the Nox gas are recycled as nitric acid, reaction performed by water wash, countercurrent of the gaseous flow plus addition of hydrogen peroxide: NOx + O 2 HNO 3 H2OH2O

47. 47 Extraction Cycle The nitric acid dissolution solution of the irradiated fuel contains the totality of uranium and plutonium as nitrate complexes (UO 2 2+, 2NO 3 - ) and (Pu 4+, 4 NO 3 - ) and the fission products. The extractant chosen, tri-butyl phosphate (TBP) is diluted to 30% by volume in an aliphatic hydrocarbon, n- dodecane, HPT or a petroleum cut. TBP presents a great affinity towards U(VI) and Pu(IV). Its affinity towards the FP and other metals is low, which brings a very selective extraction of U (VI) and Pu(IV)

48. 48 During the contact between the organic phase and the aqueous phase, the 2 following reactions take place The reactions are strongly displaced towards right (extraction) if [NO 3 - ] is high and towards the left (reextraction) if [NO 3 - ] is low The choice of HNO 3 allows to predict the direction of the solutes (U(VI) and Pu(IV)) transfer

49. 49 1 st extraction cycle The 1 st decontamination cycle (U + Pu) is presented on the next slide. The aqueous flow (U + Pu + FP) is contacted with the organic phase in a pulsed column, countercurrent flow. The first stage is hence a co-decontamination stage during which the U and Pu are removed from most of the FP The organic phase charged in U and Pu is then washed by a nitric acid solution in order to scrub from the organic phase the small quantities of FP that was entrained with U and Pu into the organic phase.

50. 50 1 st extraction cycle (2) After this first extraction column, most of the PF are separated from U and Pu and exit as raffinate. U and Pu exit the column as organic phase The organic phase is stripped of U and Pu, and the solvent can be recycled after treatment, to eliminate the degradation products H 2 MBP (monobutyl phosphate) and HDBP (dibutyl phosphate), formed by the hydrolysis and radiolysis of TBP. U and Pu prepurified exit the second column in the aqueous solution. In certain units, the implementation of the co- decontamination cycle may be complicated by the behavior of certain FP such as Zr and Tc.

51. 51 Solution U + Pu + FP Solvent HNO 3 H2OH2O Solvent Treatment Extraction U + Pu Solvent Wash Reextraction U + Pu FP U + Pu 1 st EXTRACTION CYCLE

52. 52 Feed Aqueous Solution U+Pu+ FP* Solvent Mixing U+Pu+PF* + Solvent PF* Settling U+Pu + Solvent 1 st extraction cycle (3)

53. 53 Feed Aqueous Solution U+Pu+ FP* Solvent Mixing U+Pu+PF* + Solvent PF* Settling U+Pu + Solvent Solvent Extraction Pulsed Column Feed Aqueous Phase U + Pu + FP* Perforated Plates Exit raffinate FP* Low Mixer settler Feed Solvent Compressed Air High Mixer Settler Interphase Exit Solvent Loaded U + Pu * FP : Fission Products PF* U+Pu Mixer settler U+Pu+FP* Solvent In Feed Aqueous Phase U+Pu+FP* Solvent Out Loaded with U+Pu Exit raffinate FP*

54. 54 2 nd Extraction Cycle- Splitting Stage The second step consists in separating U/Pu realized by liquid-liquid extraction (see next figure). It exploits the different redox properties of U and Pu. In the first extraction column, the co-extraction U(VI)/Pu(IV) is again performed. After washing, the organic phase loaded in U(VI)/Pu(IV) is injected into the second column where it is contacted with an aqueous solution containing a specific Pu reducer. U/Pu splitting takes place by reductive stripping of the plutonium. Indeed, Pu(IV) is reduced to Pu(III) and migrates into the aqueous phase, the organic phase containing U(VI) is treated by water (countercurrent) to realize the extraction of U(VI) into the aqueous phase. From this step, U and Pu will go in separated purification cycles (2 more cycles)

55. 55 Solution U + Pu Solvent HNO 3 Reducer Solvent Treatment Extraction U + Pu Solvent Wash Reextraction Pu Effluents Pu U + Pu SEPARATION U/Pu CYCLE Aqueous Solution Wash Reextraction U U H2OH2O

56. 56 Reducer for Pu Several Pu reducer can be used Ferrous sulfamate (Fe 2+, 2NH 2, SO 3 - ) Pu 4+ + Fe 2+ Pu 3+ + Fe 3+ Uranium (IV)* 2Pu 4+ + U 4+ + 2H 2 O2Pu 3+ + UO 2 2+ + 4 H + Hydroxylamine nitrate (NH 3 OH +, NO 3 - )* (*: the most used)

57. 57 Final conditioning Uranium A purified uranyl nitrate solution is concentrated by distillation and the concentrated solutions ( U > 400 to 500 g/L) are essentially stored waiting to be reused. For U coming from PWR, a fraction is recycled for new enrichment, which requires the steps of precipitation/filtration of ADU, thermal decomposition which gives U 3 O 8 and then transformation into UF 6

58. 58 Final conditioning (2) Uranium (2) The recycling of uranium produced by LWR fuel reprocessing is to as easy as the U-235 content of 1.07% would tend to imply. In fact, the uranium thus produced is also enriched with isotopes of atomic weight 232 (harmful due to hard g emitters decay products) and neutron absorbents U-234 and U-236. In France, however, the decision has been taken to recycle reprocessed uranium

59. 59 Final conditioning (3) Plutonium Plutonium is precipitated from the purified aqueous solution by addition of oxalic acid The plutonium oxalate, after filtration is calcinated at 450°C, which gives PuO 2

60. 60 Final conditioning (4) Plutonium(2) The plutonium recycle strategy currently adopted by many countries consists in the preparation of mixed fuels (U, Pu)O 2 (~7% Pu) called MOX for LWR reactors. A plant is build in France, MELOX, with a production capacity of 100 tons of heavy metal per year. A plant is being constructed in USA (August 2007), the MFFF facility, Savannah River Site (see class 20)

61. 61 Final conditioning (5) Military Plutonium In the case of military application of plutonium, Pu metal will be fabricated, 2 steps are necessary: Fluorination Calciothermy High Temperature

62. 62 Waste Conditioning Most of the radioactivity of the irradiated fuels can be found in the waste issued from the reprocessing plant. These waste from different flows (gas, liquid, solid) need to go towards different treatment units for –Reduction of their volume –Immobilization in specific matrices for their temporary or permanent storage to avoid their release towards the biosphere

63. 63 Gaseous Waste The different gaseous flows from the decladding unit and the off-gas from the dissolvers contain dangerous radionuclides: Kr-85, I-131(I 2 ) and Ru-104 (RuO 4 ) Furthermore considerable quantities of NOx gas are produced, that need to be recycled. Gas are washed, countercurrent wash, with water in the presence of H 2 O 2, which gives the formation of HNO 3 that can be recycled. NaOH trapping will trap I 2 and Ru, radioactive iodine can be conditioned as PbI 2, that can be safely stored. After going through different filters, the gas, that contain still radioactive isotopes of noble gas such as Kr-85, are rejected into the environment through chimneys that allow the rapid active gas dilution.

64. 64 Liquid Waste The essential of FP is in the raffinate of the 1 st extraction cycle. The other radioactive effluents produced by the other purification cycles contain lower amount of radioactivity. A second series of liquid effluents is made of degraded organic phases We will see in class 21 the treatment of these effluents

65. 65 Solid Waste 2 categories Waste issued from the process –Cladding –Gas Traps (PbI 2, AgI) –Muds produced during the treatment of liquid effluents –Solid products issued form the final treatment of the solvent –Ion exchange resins ( treatment of the pools) Technological waste –Apparatus –Pumps…

66. 66 Solid Waste (2) The main solid residue issued from dissolution consist of sections of clad or hulls weighing 300 kg per ton of fuel. For the time being, these wastes are stored at La Hague under water, amounting to interim storage. In the UP 3 and UP2 800, La Hague France, the hulls are immobilized in a concrete matrix for subsequent final disposal.

67. 67 Solid Waste (3) The different waste are classified as a function of the radioactive type and its importance ( , , , low, medium and high level waste). The conditioning is specific, it can be Cementation (PWR hulls) Wrapping in Epoxy resin (for the ion exchange resins) Bitumen isolation (mud from the effluent treatment unit)